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A high pressure water loop was constructed to investigate thermal-hydraulic characteristics under operational and transient conditions of a high conversion light water reactor (HCLWR). The maximum pressure, heating power and heating length of the loop are 16MPa, 1.1MW, and 1.8m, respectively. The test section consists of seven electrically heated rods arranged in a triangular lattice. Axial power distribution is simulated stepwisely under the maximum linear power density of 75kW/m. Several C-A thermocouples are embedded inside the cladding to detect the onset of departure from nucleate boiling (DNB). The heating power of the center rod is about 17% higher than that of the peripheral rods to ensure the first DNB detection at the center rod. Flow and power control system has been developed to perform accident simulation tests and transient DNB tests. Accident simulation tests of a double-flat-core type HCLWR were performed with the loop. Among various accidents, locked rotor of one primary coolant pump and control rod cluster ejection were selected as typical severe events caused by flow reduction and power increase, respectively. After several trials, the transients of flow rate and surface heat flux could be well agreed with the core thermal-hydraulic behaviors calculated with the three-dimensional best-estimate code REFLA/TRAC. DNB was not observed in the course of the transients as in the case of the analyses. Transient DNB tests were then performed by increasing the initial heating power while preserving the power transient curves of the accident simulation tests. The test results indicated that the double-flat-core type HCLWR has a large DNB margin under these two accident situations. The local heat flux at the onset of DNB was predicted within the uncertainty of 10% by either KfK or EPRI-Columbia critical heat flux (CHF) correlations using local instantaneous flow conditions calculated with the COBRA-IV-I subchannel analysis code. (J.P.N.). (ERA citation 17:028339)